The invention disclosed herein relates to nozzles, nozzle assemblies, and nozzle repair assemblies for vessels and piping that are installed either initially or as replacements without any welding at all to the vessel. (A "nozzle" may be, or include as part thereof, a sleeve and/or, piping. A "vessel" may also be large bore piping.) The invention more particularly relates to nozzles, nozzle assemblies, nozzle repair assemblies and procedures which replace or repair nozzles that are attached to the vessel on the inside diameter of the vessel with a J groove structural weld. (The term "nozzle assembly" is sometimes interchangeably used herein with the term "partial nozzle assembly" which encompasses an assembly that includes part of an existing nozzle and a partial replacement nozzle or the term "full nozzle assembly" which encompasses an assembly that includes a full replacement nozzle. Both partial and full nozzle assemblies in addition may include components such as gaskets, packing glands, a spring, and anti-rotation and leak path devices etc. These assemblies provide a mechanical attachment of a nozzle to a vessel, a mechanical seal of a nozzle, or both. The term "nozzle repair assembly" encompasses an existing nozzle and a structure which mechanically seals the existing nozzle or attaches the existing nozzle to the vessel without welding to the vessel, or both.) The invention has particular application to nozzles, nozzle assemblies, nozzle repair assemblies and procedures which replace, repair or initially install nozzles in ASME pressure vessels and large bore piping of pressurized water reactor (PWR) nuclear power facilities, either because a nozzle has failed, or to prevent future failure or leakage due to a phenomenon known as Primary Water Stress Corrosion Cracking, PWSCC. Repair of a nozzle also encompasses plugging the bore, for example, replacing a heater with a plug. Further, the invention relates to replacement and repair of nozzles and assemblies that were previously replaced or repaired.
A typical PWR nuclear power generating facility includes in part a reactor vessel, steam generator, pressurizer vessel, and a reactor coolant piping system, all of which operate under high pressure. Nozzles are attached to the vessels and/or piping for a number of purposes, e.g., for connecting piping and instrumentation, vents, and to secure control element drive mechanisms and heater elements. A typical pressurizer vessel 20 is shown in FIG. 1 with nozzles 22 for vents, nozzles 24 for sample or liquid level, nozzles 25 (one of which is shown behind nozzle 24) for pressure sensing, a nozzle 26 for temperature measuring, and a number of nozzles 27 for heating elements. All of those nozzles were heretofore welded to the pressurizer vessel at the time of original manufacture.
As shown in FIG. 2, inconel or stainless cladding 29 is welded to the interior of the pressurizer vessel which is made of carbon steel. The nozzle 26 shown in 5 cross section in FIG. 2, which is exemplary of the welded nozzles 22 and 24-27, passes through a hole or bore 30 in the pressurizer vessel 20 and is structurally welded at its interior end 32 to the vessel 20 with a J-groove weld 34 along the interior opening to the bore 30. The diameter of nozzle 32 is slightly less than the diameter of bore 30, so that there is a small annular space 36 between the nozzle exterior and the wall of bore 30. In some applications the nozzles are fit tight to the bore, and in a control rod drive mechanism, they are installed with a shrink fit process. The J-groove weld 34 also functions as a seal weld to seal the annular space 36. A reactor vessel (not shown) similarly has nozzles represented by nozzle 26 in FIG. 2 welded thereto. The piping of the reactor coolant system (not shown) also includes similar nozzles welded thereto. Further details of pressurizer vessels, reactor vessels, and coolant system piping, in particular, and nuclear power facilities, in general, are known to those of skill in the art.
As mentioned, the invention has particular application to the replacement of nozzles which have failed, and to the prevention of nozzle failures and leakage in nuclear power facilities due to the PWSCC phenomenon, which occurs on components having a susceptible material, high tensile stresses, high temperature and which are in a corrosive environment, conditions which primarily exist on nozzle penetrations in the pressurizer vessel, reactor coolant piping, and the reactor vessel. Such failures are manifested by cracking, which the applicant recognized resulted from several contributing factors. Such cracking occurs at the grain boundaries on the inside diameter of the nozzle material (Alloy 600) at or near the heat affected zone of the weld and propagates radially outward through the thickness of the nozzle which eventually leads to small leakage of the reactor coolant supply. Recently, failures have also occurred on stainless steel pressurizer nozzles though the failure mechanism is not known.
As indicated, nozzles of these types have failed over time and have had to be replaced or repaired, either because of a failure in the nozzle or the weld attaching and sealing the nozzle to the vessel. A typical replacement procedure in a nuclear power plant environment requires shutting down the nuclear power plant, removing the nozzle, which typically requires machining operations, and welding a replacement nozzle to the vessel or piping. The welded replacement nozzles currently in use closely duplicate the original welded nozzle they replace, except that they may be made of a different alloy, e.g., Alloy 690 (less susceptible to PWSCC) instead of Alloy 600, and may also be represented by the nozzle shown in FIG. 2. Other weld repair methods involve installing a thick weld pad on the outside of the vessel and structurally welding the nozzle to the pad, and seal welding the interior end of the nozzle to the vessel. Other refinements to this method involve inserting a thin sleeve between the nozzle and vessel, seal welding the thin sleeve on the inside diameter of the vessel, and structurally welding the thin sleeve and nozzle on the outside of the vessel. This methodology promotes less weld residual stress and eliminates the axial thermal stresses imposed since the nozzle is now free ended and allowed to grow thermally when the vessel heats up. Still another derivative of the weld pad designs is the use of a partial nozzle instead of a full nozzle. With this approach, only a portion of the nozzle is removed leaving the cracked portion of the nozzle remaining in place. A partial nozzle is structurally welded to the weld pad leaving approximately a 1/16" to 1/8" gap between the existing and new nozzle. A final derivative of the weld pad is replacing a portion of the existing heater nozzle, otherwise known as a heater sleeve, with a plug instead of a partial nozzle or sleeve, which is used due to the inherent difficulties of aligning the replacement nozzle such that the heaters can be reinstalled. Therefore, the failed heater location is abandoned in place with the plug preventing further leakage. Also, a heater sleeve may be plugged because a replacement for the failed heater is not available during the repair period, typically a scheduled outage. The heater will then be replaced during next scheduled outage.
The above partial nozzle replacement and plugs are normally used as last resorts since the carbon steel vessel is exposed to reactor coolant containing boric acid which can corrode the vessel and since the failed nozzle can continue to crack with the crack propagating into the base material of the vessel. Boric acid corrosion, though a concern, has not proven to corrode the vessel bore to any sufficient depth for at least one nozzle installed in the steam space in the pressurizer, which is in a low oxygen, stagnant environment. However, corrosion could be a concern for those areas with high fluid flow such as nozzles on reactor coolant piping.
Another method of plugging a heater location, though currently not used in the industry, involves removing a portion of the existing nozzle, inserting a new nozzle with split flared ends that compress during insertion and expand once the nozzle is fully inserted such that the flared ends overlap the inboard end of the existing nozzle relying on the integrity of the existing nozzle and weld to prevent ejection of the replaced nozzle. Once the split flared nozzle is installed, a portion of the nozzle is then rolled with high compressive load expanding the rolled area into the vessel bore, thereby forming a seal therebetween. After the rolled joint is formed a steel rod is inserted through the nozzle past the split flared ends, which prevents them from compressing inward, and a seal weld is applied at the outside diameter of the vessel at the nozzle and rod interface. The inherent problem with this plug is the rolled joint. The rolling process causes a compressive stress layer in the joint itself, but at the interface between the rolled and non rolled area, high tensile stresses are imposed which promote cracking. Cracking has been observed in the rolled transition region on similar repairs used for steam generator A600 tubes. Though this method uses A690 as the replacement material, A690 is not believed to be immune from PWSCC. Furthermore, the rolled joint may not be reliable to form a seal for extended periods because of relaxation over time in the rolled area. Any leakage here would cause severe corrosion to the outside of the vessel because as the reactor coolant leaks to the atmosphere, it flashes to steam and is in a highly oxygenated environment.
Replacements employing the above-described procedures in a nuclear power plant currently require a minimum of approximately fourteen days for some types of nozzles (although some claim a minimum of seven to ten days), and are extremely expensive. Including the lost revenue resulting from plant shut-down, which may be as high as $750,000 per day, the total cost of each repair can result in several millions dollars if the repair has to be performed in a non-scheduled facility shut down.
The above-described nozzle replacement and repair procedures and any other replacement and repair procedure that requires welding to the vessel not only is time consuming and therefore expensive, but also exposes repair personnel to more radiation than the replacement and repair procedures disclosed herein, particularly where the nozzle replacement or repair procedure involves personnel entering inside the vessel to perform the procedure. Also, both the original welded nozzle and the known welded replacement nozzles and methods subject the welded nozzle to high residual stresses imposed by weld shrinkage. These high residual stresses increase the susceptibility to PWSCC. Thus, the welded replacement nozzle offers no improvement over the original nozzle in terms of expected life and reduction of failures, other than any improvement that may result from use of a superior nozzle material. Although, Alloy 690 material is less susceptible to PWSCC than Alloy 600, it is not known that the change in nozzle material alone will not eliminate the possibility of nozzle failures. At least three utilities have replaced nozzles using the original design criteria, Alloy 690 material for the nozzle, and alloy 600 weld material, and have experienced failures in the weld material itself Based on this information, improved nozzle replacement and repair methods are needed.
There now appears to be some recognition in the field that replacement and/or repair of nozzles in pressure vessels of nuclear power plants may proceed without structurally welding a replacement nozzle to the vessel. For example, U.S. Pat. Nos. 5,149,490 and 5,202,082 (both of Brown et al.) and U.S. Pat. No. 5,404,382 (Russ et al.) (see FIGS. 8 and 9) describe methods and apparatus for replacing a nozzle for a pressurizer vessel without structurally welding the nozzle to the vessel. Although the replacement nozzles of the '490 and '382 patents are mechanically attached to the pressurizer vessel, according to these patents welding is still required to provide the seal between the nozzle and the pressurizer vessel. Therefore, the residual stresses discussed above are imposed on the nozzle by the weld whether it be a structural weld or a seal weld, which imposes less stress than a structural weld. The '382 patent discloses with respect to FIGS. 8 and 9 that the weld seams are free from loading because the nozzle is prestressed in a such a way that axial and horizontal forces acting on the nozzle are compensated for. Here, as understood by the applicant, the purpose of the prestressing is to compress the tapered nozzle against the vessel bore to eliminate lateral movement of the nozzle and bending at the weld location. It appears that the replacement procedure disclosed in the '382 patent can only be used in those locations where the inside of the vessel can be accessed such as the reactor head. This method cannot be used in the heater locations and most likely could not be used in the reactor coolant piping (typically 28" piping located in a confined area, with a high dose rate that can not be easily shielded).
In the replacement procedure and nozzle described in the '082 patent, the original welded nozzle is not fully removed, and a mechanical seal is made between the remaining cracked nozzle portion and the end of the replacement nozzle. Leaving part of the existing nozzle at the interior welded may lead to future cracking in the existing failed portion of the Alloy 600 nozzle that was not removed from the vessel. However, further cracking in the base material of the vessel would not likely result in reactor coolant leakage. Experience and analysis indicate that the base material of the vessel will only crack some minimum distance into the base material which will not jeopardize the integrity of the vessel. Boric acid corrosion for this design is less of a concern than the partial nozzle welded design discussed above due to the fact that there is not an exposed gap between the new and existing nozzle thereby limiting or eliminating boric acid from contacting the vessel. Partial nozzle designs not having sealing capabilities between the new and existing nozzle can be justified readily where the nozzles are located in stagnant region of the pressurizer vessel, e.g. the liquid space or the upper steam space because there is a limited amount of bore cavity available for the stagnant boric acid to collect and oxidize between fuel cycles and because the reactant coolant has a low oxygen content. Further, if corrosion does occur, a protective oxidation layer will form which resists further corrosion in the stagnant environment. Therefore, partial nozzle replacements without a seal between the existing nozzle and the partial replacement nozzle appear to be justified for the life of the plant. For those locations in which the partial nozzle replacement without the seal is not in a stagnant environment, i.e., reactor coolant piping, additional evaluations are required to justify extended usage. Partial nozzle designs with the seal between the existing nozzle and the partial replacement nozzle can be used in stagnant environment locations and at high fluid flow locations; however, a better design practice would be to remove the cracked nozzle to eliminate further degradation of the vessel. The procedure described in the '082 patent thus has the drawback that a portion of the failed nozzle remains structurally welded to the vessel and therefore continues to subject the vessel to the same stresses as the original nozzle, although, as mentioned above, there now appears to be justification for employing partial nozzle replacements. In any event, the remaining nozzle portion and the vessel portion surrounding the bore opening are subject to further degradation.
As far as the applicant is aware, the replacement nozzles and replacement procedures disclosed in the '490 and '082 patents have not been used in a nuclear power facility anywhere, and the replacement nozzles and nozzle replacement procedures disclosed in FIGS. 8 and 9 of the '382 patent have not been used in a nuclear power facility in the United States.
A repair procedure has also been developed which does not use welded replacements and plugs. This repair procedure involves leaving the existing nozzle installed, installing a split flange exterior to the vessel, applying a split seal at the outside diameter of the vessel and nozzle interface, and providing an unloaded anti-ejection device. In the event the existing nozzle were to crack due to PWSCC, the seal at the outside of the vessel is intended to prevent reactor coolant leakage. This repair procedure is intended as a preventative leakage measure as opposed to an ASME code section XI type of repair or replacement. Nuclear Regulatory approval may be required before this repair procedure could be used. However, it has already been installed at one utility as a preventive measure. This methodology does not prevent PWSCC from occurring. However, without further evaluation it is unknown by the applicant if the repair is acceptable for the life of the plant. It is possible that if the nozzle began to crack at numerous locations, the crack could propagate axially from the J-groove weld to beyond the seal incorporated at the flange. However, if this occurred it would not be a significant safety concern but would require additional repairs.
The following U.S. patents relate to installing, replacing, repairing or sealing nozzles, sleeves, tubes or plugs in pressure vessels and piping: U.S. Pat. No. 2,711,841 (Schaefer et al.); U.S. Pat. No. 4,255,840 (Loch et al.); U.S. Pat. No. 4,440,339 (Tamai et al.); U.S. Pat. No. 4,480,841 (Schukei et al.); U.S. Pat. No. 4,615,477 (Spada et al.); U.S. Pat. No. 4,653,483 (Margotta); U.S. Pat. No. 4,723,795 (Shenoy); U.S. Pat. No. 4,826,217 (Guerrero); U.S. Pat. No. 5,091,140 (Dixon et al.); U.S. Pat. No. 5,094,801 (Dixon et al.); U.S. Pat. No. 5,196,160 (Porowski); U.S. Pat. No. 5,209,895 (Wivagg); U.S. Pat. No. 5,271,048 (Behake et al.); U.S. Pat. No. 5,274,683 (Broda et al.); U.S. Pat. No. 5,278,878 (Porowski); U.S. Pat. No. 5,297,187 (Sodergren et al.) U.S. Pat. No. 5,323,428 (Porter et al.); and U.S. Pat. No. 5,367,768 (Weems).
Also a method similar to that used in steam generator tube repairs has been proposed with certain modifications to the Nuclear Regulatory Committee for repairing a leaking nozzle. According to the proposal, the existing nozzle is rolled and deformed against the ID of the vessel such that a seal is created between the nozzle and vessel. (See U.S. Pat. No. 5,367,768 (Weems), cited above). Also, an optional A690 sleeve could be rolled into the existing nozzle forming a seal therebetween. A similar design was also proposed for a plug. However, the Nuclear Regulatory Committee declined the proposals because that rolling technique causes high tensile stresses at the rolled transition region which promotes PWSCC, and because that repair method may be only leak limiting which could allow the boric acid in the reactor coolant to erode a portion of the carbon steel vessel.
Nozzles are currently being replaced or repaired in PWR nuclear power facilities both because they have failed and as a preventive measure where a statistical analysis has indicated a high probability of a future failure. Nozzle failures and such statistically indicated failures have been occurring frequently enough to be a major concern for nuclear power plant operators (and owners) for a number of reasons including the high cost of repairs and the millions of dollars in lost revenue due to plant shut down. Therefore, there is a need for procedures for replacing and repairing nozzles that have failed or may fail or leak in the future, that (a) reduce the time and expense required to make the replacement or repair and (b) do not require confined entry into a pressure vessel, which reduce radiation exposure to the personnel performing the replacement or repair, and (c) reduce the susceptibility to PWSCC and do not result in further degradation of the vessel, and accordingly reduce the risk of future failures and leaks. A similar need also exists for a nozzle for initial installation applications and a method of initially installing such a nozzle in a vessel.
The invention disclosed herein addresses the above-described needs and avoids the problems discussed above, and provides original nozzles and nozzle assemblies, replacement nozzles and nozzle assemblies, nozzle repair assemblies and procedures for installing and repairing nozzles and nozzle assemblies mechanically in pressure vessels in nuclear power facilities (and in other fields) that avoid some or all of the problems discussed above, and which do not employ a structural weld or a weld of any kind to the vessel.